1. Field of the Invention
This invention relates generally to boiling water nuclear reactors (BWRs) and more particularly to nuclear fuel assemblies and the fuel rods used in BWRs for electrical power generation.
2. Description of Related Art
The generation of heat energy through fission of nuclear fuel in a nuclear reactor is well known. The nuclear fuel is located in the core of the reactor. Typically, the core of a light-water reactor (LWR) contains a plurality of fuel assemblies that each contain a plurality of fuel rods. The fuel rods are arranged with their axes along the vertical direction. Typically, a fuel rod consists of a metal cylindrical cladding that contains a stack of fuel pellets. The fuel pellet material, the size of the fuel pellets, the cladding material and the orientation of the fuel pellets within the fuel rod are all well known for commercial LWRs used in electric power production.
The radial spacing between fuel rods is filled with water, which flows in the axial direction from the bottom of the core to its top. The water carries away heat from the reactor core that is generated by fission reactions within the fuel rods. Hence, the water acts as the core coolant. The water also slows down, i.e., moderates, neutrons which are emitted at high velocities from the fission reactions. Hence, the water acts also as a neutron moderator.
The moderation of high velocity fission neutrons occurs as a result, primarily, of collisions between the neutrons and the nuclei of hydrogen atoms in the water. Consequently, the degree of moderation is a function of the number of hydrogen atoms per unit volume, for example, per cubic centimeter. As the water is heated, the water expands and so the number of hydrogen atoms per cubic centimeter decreases. When the water boils and is converted to steam, the number of hydrogen atoms per cubic centimeter decreases even further. Consequently, a region of a reactor core in which the steam occupies a significant fraction of the volume available for the coolant is undermoderated. That is, neutrons in such a region are not sufficiently moderated.
The moderation of a fission neutron, i.e, the lowering of the energy of the fission neutron, increases the probability that the neutron is absorbed in the nuclear fuel so as to cause another fission reaction which in turn releases additional high energy neutrons. When neutrons from one fission reaction cause, on the average, one other fission reaction to occur, a chain reaction is established. To generate power, a chain reaction must be established and maintained in a LWR.
The probability for achieving a chain reaction is sometimes measured by a multiplication constant and is denoted by K. In LWRs, the neutron moderation provided by the coolant makes it possible to obtain a nuclear chain-reaction, i.e., a K of one, using nuclear fuel having a relatively small concentration of a fissile isotope.
Typical LWR fuel contains uranium in which the concentration of the fissile isotope uranium-235 is enriched from the 0.71% found in natural uranium to approximately 3%. The remaining 97% of the uranium is the non-fissile isotope uranium-238. One objective in nuclear power reactor design is to lower the fuel enrichment because this reduces the nuclear fuel cost and enhances uranium resource utilization.
The enrichment of the fuel loaded into the core is chosen so that the multiplication constant K is sufficiently higher than one so that the LWR can be operated for a significant period of time, usually about 12 to 18 months, without refueling. During this period of operation, sometimes referred to as a cycle, multiplication constant K is reduced to or near one as a result of depletion, i.e., burnup, of the fissile fuel and the accumulation of neutron absorbing fission products. The deviation of multiplication constant K from unity at any instant during the cycle, commonly referred to as reactivity and equal to (K-1)/K, is offset by insertion into the core of strong neutron absorbing materials. The strong neutron absorbing materials are introduced into the core in a variety of ways, e.g., via the insertion of control rods, via inclusion of burnable poisons in the fuel, via addition of boron to the water, or via a combination of these measures.
The rate at which heat can be generated per unit length of a fuel rod, sometimes referred to as the linear-heat-rate, is limited by the ability of the water to carry this heat out of the core without having the fuel temperature and the cladding temperature exceed predetermined permissible values. The maximum permissible linear-heat-rate translates, in a given core, to the maximum power density, or power generated per unit volume of the core at the specific location under consideration.
The power density in a LWR varies across the core both axially and radially. The power density tends to drop at the core periphery because there is an increased probability that fission neutrons will leak out of the core without causing another fission reaction. The power density also tends to drop in the vicinity of control rods. The control rods, which are used to regulate the chain-reaction, contain materials, such as boron, that have a large capture probability for moderated neutrons. The power density is usually depressed also in core regions which are undermoderated relative to other core regions.
The depressions in power density are undesirable except, possibly, near the core periphery. The flatter the power density, i.e, the more the power density is constant across the core, the smaller the size of the core required to generate a given amount of electricity and consequently the more economical the reactor. Alternatively, the flatter the power density of a core of a given volume and power, the longer the residence time of the fuel in the reactor which also improves the reactor economics. The advantages of a flat power distribution have been widely recognized, but such distributions are limited by other factors, which are described more completely below.
FIG. 1 illustrates a cross-sectional view of a portion of a typical boiling water reactor BWR core 20 containing a plurality of a rectilinear fuel assemblies 12 and a plurality of control rods 24. (A boiling water reactor is one type of a LWR.) Each fuel assembly 12 includes a square channel 21 that encloses an array of fuel rods and water rods.
A gap 22 exists between all fuel assemblies 12. Gap 22 is filled with water and is called a "water gap." Gap 22 is used for the insertion of control rods 24, when necessary. The non-specified dimensions, materials, and other parameters, components and instruments associated with or part of the BWR of these and subsequent figures are conventional and are well-known to those skilled in the art.
Examples of typical BWR fuel assemblies are depicted and described in U.S. Pat. No. 3,350,275, entitled "Reactor Fuel Assembly Device" issued to Venier et al. on Oct. 31, 1967; U.S. Pat. No. 3,466,226, entitled "Nuclear Fuel Element" issued to Lass on Sep. 9, 1969; U.S. Pat. No. 3,802,995, entitled "Nuclear Fuel assembly" issued to Fritz et al. on Apr. 9, 1974; and U.S. Pat. No. 4,664,882, entitled "Segmented Fuel and Moderator Rod" issued to Doshi on May 12, 1987, all of which are incorporated herein by reference in their entirety.
Typical BWR fuel assemblies include a 7.times.7, an 8.times.8, a 9.times.9, or a 10.times.10 array of rods. For example, in an 8.times.8 array, the fuel assembly contains 62 fuel rods and two water rods. See for example, U.S. Pat. No. 3,802,995, entitled "Nuclear Fuel Assembly" issued to Fritz et al. on Apr. 9, 1974. A primary purpose of the water rods, which are typically in the interior of the fuel assembly, is to provide additional moderation.
BWR fuel assembly 12 may also include all fuel rods. A drawback of a BWR fuel assembly with all fuel rods is that the central region of the fuel assembly is undermoderated relative to the fuel assembly periphery, as the water of water gaps 24 contributes to neutron moderation near this periphery. One consequence of this undermoderation is that the power density at the central part of the assembly is lower than at the periphery of the assembly. The resulting nonuniform radial power distribution across the fuel assemblies has a negative effect on the BWR economics. This is so because such a power distribution limits the total power output and the uranium fuel utilization of fuel assembly 12.
While the water rods mitigate the undermoderation in the central region of fuel assembly 12, the water rods reduce the amount of fuel per assembly. Consequently, the total power that can be obtained from the BWR core and the fuel residence time in the BWR is limited by either the water rods or the nonflattened radial power distribution in a fuel assembly when the water rods are eliminated.
Weitzberg proposed in U.S. Pat. No. 4,591,479, entitled "Boiling Water Reactor Fuel Bundle," and issued on May 27, 1986 to improve the moderation in the central region of a BWR fuel assembly by replacing fuel pellets in a certain number of fuel rods in the center region of the assembly by pellets made of zirconium hydride. However, this approach is less effective and more expensive than the use of water rods.
In a BWR, the coolant, i.e., water, flows axially from the bottom of the core to the top of the core. As the water moves through the core, the water is heated by the heat generated in the fuel rods. In fact, the water starts to boil and to create steam before leaving the core. From the boiling initiation point upwards, the amount of liquid decreases and the amount of vapor, i.e., steam, increases. The larger the volume fraction occupied by the steam, the smaller becomes the number of hydrogen atoms per unit volume and the smaller becomes the moderation capability of the water. Thus, BWR cores are highly undermoderated in the upper region where steam is formed and the undermoderation increases as the volume of steam increases higher in the core region.
As a result of the undermoderation in the upper regions of the core, the fission rate and, hence, the axial power density is highly asymmetric unless measures are taken to flatten the axial power distribution. A typical axial power distribution is illustrated in FIG. 2. The axial power density tends to peak in the lower part of the core and to strongly decline near the top of the core.
If uncorrected, this non-uniform axial power shape can limit the overall reactor power output, as the peak power density can not exceed a given limit. The non-uniform axial power shape results in an uneven axial burn-up of the fuel. At the beginning-of-life of the fuel assembly, the rate of fuel consumption at the lower part of the assembly is significantly higher than at the upper part of the assembly. This non-uniform fuel consumption impairs overall fuel utilization.
Moreover, near the end of a cycle, the depleted fuel at the bottom core results in a lower fission rate which in turn reduces the amount of heat transferred into the water at the lower part of the core. Thus, with time the onset of boiling moves higher up the core. As the steam volume fraction decreases, the reactivity in the upper part of the core becomes significantly higher than the reactivity at the lower part of the core. This reactivity imbalance can impair reactor safety as it reduces the effectiveness of the control and safety rods, which in a BWR enter the core from its bottom. Specifically, the highly non-uniform fuel consumption can reduce the cold shutdown reactivity margin as well as prolong the time it takes to quickly shut-down, i.e., scram, the reactor.
Another drawback of the undermoderation in the upper region of a BWR core is that a significant fraction of the fission neutrons born in the region leak out of the core before being moderated. These relatively energetic neutrons damage components located in the vicinity of the core.
A number of measures are presently being used in boiling water reactors to modify the axial power shape so as to minimize the disadvantages of the asymmetric axial power distribution. These measures include the use of control rods and of consumable neutron absorbing materials, also known as burnable poisons.
Burnable poisons, such as gadolinium, are generally incorporated within the fuel rod unevenly, with the highest concentration in the lower part of the rod. Having a large probability of absorbing moderated neutrons, the burnable poison reduces the fission probability and, hence, suppresses the fission density. In addition, control rods are partially inserted into the core from its bottom, thereby absorbing moderated neutrons and thus suppressing fissions in this part of the core.
Both of these methods of axial power shaping have an adverse effect on the fuel utilization in a BWR. If the number of neutrons absorbed in the burnable poisons and in the control rods were reduced, either the enrichment of the uranium loaded into the BWR could be reduced or the fuel residence time and burnup in the reactor could be increased. Both of these changes could improve the BWR economics. Moreover, both of these methods of axial power shaping do not eliminate the safety and neutron damage drawbacks resulting from the significant undermoderation in the upper part of the BWR core.
A number of design modifications were proposed for eliminating or reducing the drawbacks associated with undermoderation of the upper part of a BWR core. In U.S. Pat. No. 3,145,149, entitled "Boiling Nuclear Reactor and Fuel Element Therefor" and issued Aug. 18, 1964, Imhoff suggested fuel rod designs in which the average quantity of fuel per unit length of fuel rod decreases towards the top of the core. Although this design reduces the degree of undermoderation in a BWR, the design significantly complicates the fuel rod fabrication. Furthermore, this design raises safety concerns. For example, fuel from upper pellets may disintegrate, fall through the central gap, and accumulate in a lower part of the fuel rod. This might lead to fuel and/or cladding meltdown which is a very undesirable accident.
Gylfe, in U.S. Pat. No. 3,145,150, entitled "Fuel-Moderator Element for a Nuclear Reactor and Method of Making" and issued Aug. 18, 1964, proposed to use fuel rods having a double wall cladding. The inner clad was a tube made of a hydride material, e.g., zirconium hydride, that was filled with fuel pellets and cladded on the outside, with a thin layer of stainless steel or another material. As the number of hydrogen atoms per unit volume of zirconium hydride is comparable to hydrogen density in liquid water at room temperature, the zirconium hydride is an effective moderator material. Thus, the zirconium hydride as the primary fuel rod cladding material reduces the level of undermoderation in the upper part of a BWR core. A drawback of this scheme is that if the zirconium hydride cladding is made thick enough so as to provide significant moderation, the resistance of the zirconium hydride cladding to heat transfer significantly increases the fuel temperature. This may limit the power which can be extracted from a fuel rod and, hence, from a core of a given size.
Weitzberg, in U.S. Pat. No. 4,591,479, entitled "Boiling Water Reactor Fuel Bundle," and issued on May 27, 1986, proposed to improve the moderation at the upper part of BWR cores by replacing fuel pellets at the upper part of a certain fraction of the fuel rods by pellets made of a solid moderator, such as zirconium hydride. Unfortunately, replacement of fuel pellets with zirconium hydride pellets reduces the amount of fuel and the total length of fuel rods in the core which in turn cancels most, if not all, of the improvement due to the increased moderation. Another drawback is that the zirconium of the zirconium hydride absorb a significant fraction of the moderated neutrons.
Uchikawa et al., in U.S. Pat. No. 4,652,427 entitled "Fuel Assembly," and issued on Mar. 24, 1987, proposed to incorporate in a BWR fuel assembly a number of rods which contain burnable poisons mixed with a solid moderating material. A drawback is that this assembly does not significantly improve the undermoderation in the boiling part of the core. Another drawback is that it limits the amount of fuel and the total length of fuel rods which can be loaded into the BWR core.
Doshi, in U.S. Pat. No. 4,664,882, entitled "Segmented Fuel and Moderator Rod," and issued on May 12, 1987, proposed to improve the moderation in the upper part of a BWR core by replacing one or more conventional fuel rods with segmented rods which contain fuel pellets in their lower part and water in their higher part. A drawback of this invention is that it reduces the amount of fuel and the total length of fuel rods in the core.
Taleyarkhan, in U.S. Pat. No. 4,818,478, entitled "BWR Fuel Assembly Mini-Bundle Having Interior Fuel Rods of Reduced Diameter," and issued on Apr. 4, 1989, proposed to improve the moderation across the BWR fuel assembly by dividing an 8.times.8 fuel rod lattice into four 4.times.4 bundles that are separated by a cross shaped water gap in between. Moreover, the four inner fuel rods of each bundle are to have a smaller diameter, so as to provide more volume for water. A drawback of Taleyarkhan's invention is that it is more complicated than present BWR fuel assembly designs. Another drawback of this invention is that it does not significantly reduce the large variation in the degree of moderation along the fuel assembly.
Typically, modern BWRs have a thermal power rating of 3000 to 4000 Megawatts, a fuel rod length of about four meters and a fuel rod outer diameter of about 1.25 centimeters (cm). The fuel, in the form of cylindrical pellets, is enclosed within a zircaloy, i.e., a zirconium alloy tube (also referred to as the cladding), nearly 0.9 millimeters (mm) in thickness. The fuel used by all BWRs is uranium oxide (UO.sub.2). In fact, with very few exceptions, oxide fuel is used in all the commercial power reactors operating around the world, including in pressurized water reactors (PWR), heavy water reactors (HWR) and even in liquid metal cooled reactors (LMR). The exceptions are a small number of gas cooled reactors (GCR) which use a metallic uranium alloy for their fuel. Metallic uranium alloy is also being considered for LMR under development in the USA. High temperature gas cooled reactors (HTGR) under development are designed to use uranium carbide and, possibly, also uranium oxide fuel.
With one exception, the fuel for reactors used for research rather than for power production is a metallic uranium alloy, uranium oxide or uranium silicide. The exception is the so called TRIGA reactor which uses a hydride of a uranium-zirconium alloy for its fuel. Typically, the TRIGA reactor fuel rods are about 30 cm in length and about 3.5 cm in outer diameter and use 0.5 mm thick stainless steel cladding. The uranium-zirconium hydride composition used for the TRIGA fuel has, typically, 1.6 hydrogen atoms per zirconium atom, denoted as U-ZrH.sub.1.6. Details about the TRIGA fuel fabrication, properties and performance can be found in many publications, such as in the General Atomics report GA-A16029 by M. T. Simnad entitled "The U-ZrHx Alloy: Its Properties and Use in TRIGA Fuel" (Aug. 1980).
The hydride fuel was selected for the TRIGA reactor primarily for its large negative effect on reactivity as the fuel temperature rises. This large temperature coefficient of reactivity enables the TRIGA reactor to generate power pulses, for the purpose of conducting various kinds of experiments. The power pulsing capability is one of the unique features of TRIGA reactors.
Compared with a BWR core, a TRIGA core is small, has a low average power output and operates at low temperatures. Hence, a TRIGA core and the conditions under which a TRIGA core is operated are different from those of a BWR core.
The small size of the TRIGA core means that larger fraction of fission neutrons leak from the core than in a BWR core. Thus, TRIGA reactor fuel needs a larger enrichment of uranium-235 to maintain the chain reaction. Traditionally, TRIGA fuel contained nearly 10 wt. % uranium that was enriched to more than 70% in U-235. Following the adoption of policies by the U.S. Government to limit the export of fuels to those enriched to less than 20% in U-235, General Atomic of San Diego, Calif. developed TRIGA fuel containing up to 45 wt. % uranium. For TRIGA and any other application of this fuel type, the uranium enrichment was 20% or very close to 20%. Whereas the former type of TRIGA fuel belongs to the category of Highly Enriched Uranium (HEU) fuel, the latter and present type of TRIGA fuel belongs to the category of Medium Enriched Uranium (MEU) fuel. Commercial light water reactors fuel use a Low Enriched Uranium (LEU) fuel.
The uranium-zirconium hydride in the composition of this enriched fuel is reported by General Atomics to be stable and operational at temperatures up to 700.degree. C. (See above cited General Atomic reference). In fact, uranium-zirconium hydride fuel rods were successfully operated with linear-heat-rates comparable to the maximum linear-heat-rate a typical BWR fuel is designed to operate.
The TRIGA core, in addition to its high enrichment and small size compared to a BWR core, is more symmetric than a BWR core. In particular, the partial water voiding associated with boiling is not encountered in a TRIGA core. Consequently, the upper part of the TRIGA core is as well moderated as its lower (or any other) part, and its power distribution is more symmetric than in a BWR core. Further, the TRIGA core is not subjected to the safety issues encountered in BWR cores due to the voiding of the upper part of these cores.
Typically, neither the TRIGA reactor, nor any other type of research or power reactor, uses a combination of hydride and non-hydride fuel materials for their fuel. Moreover, due to the low weight percent (of about 10%) and the high enrichment (above about 70%) of the HEU in the uranium-zirconium hydride fuel developed for the original TRIGA reactors and used for many years in many TRIGA reactors around the world, the uranium-zirconium hydride has been considered as inadequate as a fuel for power reactors such as BWRs. Even the MEU fuel developed in the mid-seventies for TRIGA reactors is expensive and commonly considered as inadequate as a fuel for power reactors such as BWRs. Thus, for example, Glasstone and Sesonske, in their well known "Nuclear Reactor Engineering" text and reference book (Van Nostrand Reinhold Co., Third Edition, 1981), do not even refer to uranium-zirconium hydride, or to any other hydride material as a candidate fuel material for nuclear power reactors. The types of fuel materials they refer to are metallic, oxide, carbide and nitride.